1. Field of the Invention
The present invention relates to an in-core fixed nuclear instrumentation system and a power distribution monitoring system of a reactor such as a boiling water type reactor.
2. Description of the Prior Art
A reactor, for example, a boiling water type reactor (hereinafter, referred simply to as BWR) is provided with a power distribution monitoring system which monitors a reactor operating mode and a reactor power distribution (hereinafter, in this specification, the reactor power distribution is described as "in-core power distribution", "core power distribution", or the like), as generally shown in FIG. 23 and FIG. 24.
In the BWR, as shown in FIG. 23, a reactor pressure vessel 2 is housed in a primary containment vessel 1, and a reactor core 3 is accommodated in the reactor pressure vessel 2. As shown in FIG. 24, the reactor core 3 is constructed in a manner that many fuel assemblies 4 and control rods 5 are mounted therein. In the reactor core 3, an in-core nuclear instrumentation assembly 6 is located on a position in the reactor core 3, which is surrounded by fuel assemblies 4.
As shown in FIG. 24, the in-core nuclear instrumentation assembly 6 is arranged in a corner water gap G formed by four fuel assemblies 4, and neutron detectors 8 are discretely arranged on several positions along a core axial direction in a nuclear instrumentation tube (pipe) 7.
The neutron detector 8 is a so-called fixed type, and in a boiling water type reactor (BWR), usually, four neutron detectors are discretely arranged on a fuel effective portion at equal intervals.
Further, in the nuclear instrumentation tube 7, a TIP (Traversing In-Core Probe) conduit pipe 9 is arranged, and in the TIP conduit pipe (tube) 9, one traversing neutron detector (TIP) 10 is located so as to be movable in an axial direction of the TIP conduit pipe 9. Moreover, as shown in FIG. 23, there is located a movable type neutron flux measuring system for axially continuously measuring neutron flux by means of a matrix device 11, a TIP driving device 12, a TIP control and neutron flux signal processing system 13 or the like. A reference numeral 14 denotes a penetration section, a reference number 15 denotes a valve mechanism and a reference number 16 denotes a shielding container. These neutron detectors 8 and 10, control systems such as signal processing systems 13 and 17 (will be described later) for neutron detectors 8 and 10 constitute a reactor nuclear instrumentation system 18.
On the other hand, the in-core fixed neutron detectors 8 {LPRM (local power range monitor) detector} arranged in the reactor core are divided into some groups, and then, an average signal {APRM (average power range monitor) signal} for each group is generated so that an power level of a power range of the reactor core 3 is monitored on the basis of these APRM signals. More specifically, when an abnormally transient phenomenon or accident such that neutron flux rapidly rises up happens, the LPRM detectors 8 detect the transient phenomenon and the occurrence of accident so that, according to the APRM signals generated by the detected signals of the LPRM detectors 8, a reactor safety protection system (not shown) rapidly makes a scram operation of a reactor scram system (not shown) such as a control rod driving mechanism or the like in order to prevent fuel assemblies or the reactor core from being break down. That is, the LPRM detector 8 is constituted as a part of the reactor safety protection system.
By the way, in individual in-core fixed neutron detectors 8, a sensitivity change takes place by neutron irradiation or the like. In order to calibrate a sensitivity of each neutron detector 8 for each predetermined period during operation, the TIP (traversing neutron detector) 10 is operated so as to obtain a continuous power distribution in a core axial direction, and the sensitivity change of each neutron detector 8 is corrected by means of a gain adjusting function of a neutron detector (LPRM) signal processing system 17. A detection signal S2 detected by the neutron detector 8 is processed by means of the signal processing system 17, and thereafter, is transmitted to a process computer 20 which will be described later.
In general, the BWR is provided with a process control computer 20 for monitoring an operating mode and power distribution of a nuclear (atomic) power plant. The process control computer 20 is provided with a nuclear instrumentation control system 21 for monitoring and controlling the reactor nuclear instrumentation system 18, a power distribution simulating system 22 including a physics model having three-dimensional thermal-hydraulics simulation code, and an input-output system 23. The reactor power distribution simulating system 22 is incorporated in one or plural process control computers 20 as a program.
Further, the reactor power distribution simulating system 22 includes a power distribution simulating module 24 and a power distribution learning (adaptive) module 25.
Neutron flux signal obtained by the TIP 10 of the reactor nuclear instrumentation system 18 is processed as a nuclear instrumentation signal corresponding to a core axial direction position by means of the TIP neutron flux signal processing system 13 of the reactor nuclear instrumentation system 18. Then, the nuclear instrumentation signal is read via the nuclear instrumentation control system 21 of the process control computer 20 into the power distribution simulating system 22 as a reference power distribution in a three-dimensional nuclear thermal-hydraulics simulation.
On the other hand, core state data S3 (process quantity) including a control rod pattern, a core coolant flow rate, an internal pressure of the reactor pressure vessel, flow of feed water, a temperature of feed water (a core inlet coolant temperature) and so on, which are used as various operating parameters indicative of a reactor operating mode (state) and obtained from a core state data measuring apparatus 26 as a reactor core state data measuring means, is read into a core state data processing system 27, and then, is processed so that a reactor thermal output or the like is calculated. Then, the reactor core state data S3 including the calculated reactor thermal output is transmitted to the reactor power distribution simulating system 22 via the nuclear instrumentation control system 21 of the process control computer 20.
In fact, the reactor core state data measuring apparatus 26 is composed of a plurality of monitoring devices. In addition, the reactor core state data measuring apparatus 26 is a general name of an apparatus for collecting process data of various operating parameters of the reactor, and is shown as one measuring apparatus in FIG. 23 for simplification. Moreover, the core state data processing system 27 may be used as one function of the process control computer 20.
The detection signals S2 and the core state data S3 transmitted in the aforesaid manner are transmitted to the power distribution simulating system 22 of the process control computer 20. In the power distribution simulating system 22, a core power distribution is simulated on the basis of the transmitted core state data S3 and the three-dimensional nuclear thermal-hydraulics simulation code of the power distribution simulating module 24. Further, the power distribution simulating system 22 learns a reference power distribution of the core nuclear instrumentation data by a learning function (adapting function) of the power distribution learning (adapting) module 25, and then, corrects the simulation result (core power distribution) while referring to the reference power distribution. As a result, in a power distribution predictive simulation after that, it is possible to accurately simulate a reactor power distribution.
In the conventional in-core nuclear instrumentation assembly 6, in place of the traversing neutron detector 10, as shown in a perspective view partly in section of FIG. 25, there may be provided a traversing .gamma.-ray detector 10A which is moved in a core axial direction so as to continuously measure a .gamma.-ray flux in the core axial direction. The .gamma.-ray is generated in proportional to a fission quantity in the reactor core; for this reason, the .gamma.-ray flux is measured, and whereby, it is possible to measure a fission quantity profile in the reactor core.
By using the traversing neutron detector 10 and the .gamma.-ray detector 10A, it is possible to calibrate a dispersion of each detection accuracy of each plural neutron detector 8 arranged in the core axial direction.
As described above, in the conventional reactor nuclear instrumentation system 18, in order to continuously measure a power distribution in the axial direction of the reactor core 3, the traversing neutron detector 10 and the traversing .gamma.-ray detector 10A constituting a movable type in-core instrumentation system have been used.
However, in the traversing (movable type) neutron detector 10 and the .gamma.-ray detector 10A, measurement is made in a manner of vertically moving at least one neutron detector 10 or .gamma.-ray detector 10A over the entire length (core axial length) of the core 3 in the TIP conduit tube 9 from the outside of the reactor pressure vessel 2. For this reason, there are the problems that a mechanical driving and operating mechanism for moving and operating the traversing neutron detector 10 and the .gamma.-ray detector 10A is made large and its structure is complicate so that a moving operation and maintenance of the mechanical driving and operating mechanism are troublesome. In particular, maintenance and management are required for the mechanical driving and operating mechanisms such as the detector driving device 12 for moving and operating the traversing neutron detector 10 and the traversing .gamma.-ray detector 10A, the matrix device 11 for selecting the TIP conduit tube 9, the valve mechanism 15, the shielding container 16 or the like. In addition, the traversing detectors 10 and 10A are activated; for this reason, there is the possibility that an operator (worker) which is carried out the above maintenance and management works of the traversing neutron detector 10 and the traversing .gamma.-ray detector 10A is exposed.
In view of the above problem, in the reactor nuclear instrumentation system, there is made a demand for a method of monitoring a reactor operating mode and a power distribution in a core axial direction without using a traversing type measuring (nuclear instrumentation) system.
The in-core nuclear instrumentation assembly 6 used in the conventional reactor nuclear instrumentation system is usually provided with four fixed neutron detectors 8, one traversing neutron detector (TIP) 10 or traversing .gamma.-ray detector 10A, and a hollow conduit (TIP conduit tube 9) for housing movably the traversing neutron detector (traversing .gamma.-ray detector). In place of the TIP 10, a study has been made such that a fixed .gamma.-ray heating detector is arranged in the same manner as the fixed neutron detector 8.
However, in the case where a plurality of, for example, four fixed .gamma.-ray heating detectors, are located in a core axial direction, it is impossible to measure a power on an upper and lower portions of the reactor core 3. Further, in the case of extrapolating measurement data on the upper and lower portions of the reactor core 3 from four measurement data or interpolating these four measurement data, a behavior of power distribution change differs between individual portions in the core axial direction; for this reason, a great measurement error is caused so that an accuracy becomes worse.
If only fixed measuring (nuclear instrumentation) devices are located at several positions in an axial direction in the reactor nuclear instrumentation system, a measurement error is great in a power distribution of the core axial direction. For this reason, there is a need of previously setting a great margin on a limiting condition (operational thermal limit) for a reactor operation. As a result, a degree of the margin of the reactor operation is reduced, so that there is the possibility of giving a bad influence to an availability factor of the reactor.
Moreover, in order to improve a measurement accuracy of the power distribution in the core axial direction, it is considered that a plurality of fixed .gamma.-ray heating detectors are arranged in the core axial direction. However, in this case, a number of detector signal cables is increased, and there is a limit to arrange many .gamma.-ray heating detectors in view of restriction of the number of detector connecting cables capable of passing through the nuclear instrumentation tube 7 of the in-core nuclear instrumentation assembly 6.
As disclosed in Japanese Unexamined Patent Publication No. 6-289182, there has been proposed a reactor nuclear instrumentation system in which many .gamma.-ray heating detectors (called as GT or GT detector) are arranged. However, in the reactor nuclear instrumentation system, a .gamma.-ray heating contributing range analysis and a knowledge of .gamma.-ray heating are insufficient. Since at least one of .gamma.-ray heating detectors located on upper and lower ends of the core is arranged at a position within 15 cm from the upper and lower ends of the fuel effective portion of the core axial direction, it is difficult to accurately detect each .gamma.-ray heating value on the upper and lower ends of the fuel effective portion.
In the case of measuring a power distribution of reactor core with the use of many fixed .gamma.-ray heating detectors (GT detectors), a part of these many GT detectors is arranged in the vicinity of an LPRM detector, and thereby, the following technique has been proposed; more specifically, since the GT detector has characteristic of less variation of bias, sensitivity or gain of the LPRM detector is adjusted by means of the GT detectors, or a GT assembly, which has a plurality of GT detectors arranged in an axial direction, is used as core axial direction power distributing measuring means, in place of the traversing neutron detector or the traversing gamma-ray detector.
In the .gamma.-ray heating detector (GT detector) used in the conventional reactor nuclear instrumentation system, a differential thermocouple is used for detecting a .gamma.-ray heating temperature. For this reason, the following report has been made; more specifically, almost no aged deterioration; however, a voltage output of the thermocouple with respect to a gamma heat value lowers depending upon an elapse time for unit of week or month, and a saturated phenomenon of the voltage output happens after a stay time of the GT detector in the core (mounted time of the GT detector in the core) to some degree. Therefore, with the use of a heater incorporated into a gamma-ray thermometer (GT) assembly comprising a plurality of GT detectors, a sensitivity {sensitivity coefficient (constant); a value for determining a relationship between a thermocouple output voltage of each GT detector and a .gamma.-ray heating value (unit: W/g) per unit weight} is periodically measured. Then, the measured sensitivity value is checked, and when the measured sensitivity value is changed over a constant level, it is necessary to calculate a .gamma.-ray heating value of the GT detector from a thermocouple output voltage signal with the use of a new sensitivity coefficient corresponding to the changed sensitivity value.
In this specification, the aforesaid processing, that is, a processing of measuring a sensitivity of each GT detector with the use of the heater, and in the case where a sensitivity change with respect to the measurement result exceeds a constant level, setting a new sensitivity coefficient corresponding to the changed sensitivity, is described as "sensitivity calibration processing".
Moreover, when carrying out the aforesaid sensitivity calibration processing, a GT signal outputted from the GT detector is bypassed so as not to be used for power distribution measurement processing. In addition, the GT detector or GT assembly, which is not used for the aforesaid power distribution measurement processing because the sensitivity of the GT detector or GT assembly is being calibrated or the sensitivity of that shows a defective value so that the GT detector or GT assembly is out of order, is called as bypassed GT detector or bypassed GT assembly.
By the way, the GT assembly is incorporated in the same in-core nuclear instrumentation tube integrally with the LPRM detector assembly which are thermal neutron detectors. A sensitivity with respect to a thermal neutron of the LPRM detector is determined depending upon a change by an in-core irradiation quantity of U235 and U234 coated onto an inner surface of a fission detector. If the sensitivity with respect to a thermal neutron of the LPRM detector gets to be a constant value or less, the in-core nuclear instrumentation assembly, that is, the in-core nuclear instrumentation tube including the LPRM detectors and the GT assembly having a sensitivity lowering to a constant value or less, is replaced together. Therefore, in actual use, an in-core mounted elapse time of the gamma-ray thermometer (GT) assembly differs for each in-core nuclear instrumentation assembly.
An actual sensitivity of the GT detector in the output voltage sensitivity calibration processing by a heater incorporated into the GT assembly, is measured by the following equation (2) which will be described later, on the basis of an increase of the thermocouple voltage signal by a additional heating value of the GT detector.
For this reason, in a thermal equilibrium state that additional heating by a built-in heater is sufficiently completed, an average value must be obtained from large number of time series data of one GT detector signal. Therefore, a time of approximately 30 to 60 seconds per GT assembly is required to collect the GT signal (output voltage signal).
In an ABWR (Advanced Boiling Water Reactor) in a 135 ten-thousand kWe range, the reactor core is provided with 52 in-core nuclear instrumentation tubes each including the aforesaid GT assembly. Therefore, in the case of carrying out a calibration by the built-in heater of the GT assembly, if three circuits are prepared for each tube, about 9 to 20 minutes are required depending upon the number of power supply circuits of the heater and heating value measuring circuits of the heater.
Moreover, in a core mounted lifetime (approximately 7 years) of the GT assembly, if a heater calibration of the GT assembly is carried out according to heating by the heater, there is the possibility that the heater is latently break down. Thus, by avoiding unnecessary calibration by heater heating, and shortening a time spent for the calibration by heater heating, it is desirable to reduce a time inoperable of measuring a power distribution of the whole core by the GT signal as much as possible.
On the other hand, during a calibration of the GT assembly by heater heating and in-core power distribution measurement by the calibrated GT assembly, it is necessary that the core or in-core power distribution is a steady state for a predetermined time or more (approximately one hour when a gamma decay chain becomes a substantially equilibrium state) on the basis of the principle of measuring a gamma-ray heating value.
In the BWR, the process control computer has a built-in three-dimensional simulator, and a core power distribution simulation is periodically or always carried out with the use of the parameter of core state data such as a reactor pressure, a core heat output, a core coolant flow rate, a control rod pattern or the like, and thus, it is confirmed that a fuel assembly satisfies a core operational thermal condition (limit).
Before a time (within about one hour) relatively shorter than the point of time of a periodically core power distribution simulation, for example, in the case where the core power distribution varies by a change of the control rod pattern or a great change of the core coolant flow rate, the LPRM detector can instantaneously output neutron flux signal corresponding to a power distribution change. However, a signal (GT signal) of the GT detector becomes a precise signal level after a predetermined time, for example, one hour or more elapses because a delayed gamma-ray source slowly varies.
Therefore, it is impossible to carry out the in-core power distribution adaptive correction processing or LPRM detector sensitivity and gain adjustment processing until the GT signal becomes the precise signal level. For this reason, the in-core power distribution adaptive correction processing or LPRM detector sensitivity and gain adjustment processing can not be periodically or always carried out.